A nuclear reactor installation of the pressurized-water reactor type includes a main pressurized-water coolant pipe line loop circulating the coolant through the reactor and a steam generator producing steam representing useful power. To provide the necessary pressure on the coolant circulating in the loop, a pressurizer is connected with the loop, the pressurizer chamber receiving a portion of the coolant and having a steam space thereabove containing steam applying the pressure to the coolant.
A portion of the coolant in the loop is continuously removed and subjected to treatment and returned to the loop and in the loop the coolant may be admixed with a solution of boric acid in water as suggested by the text VGB-Kernkraftwerks-Seminar 1970, p. 41, FIG. 2. The boric acid solution there referred to has a concentration that is relatively low, being not more than 1,200 ppm, and the volume fed to the coolant is relatively small compared to the overall volume of the coolant. This practice is called chemical trimming, is used only for the normal operation of the reactor and has no appreciable effect insofar as substantial changes in the reactivity of the reactor core are concerned.
If for any reason the pressure of the coolant circulating in the loop drops, such as in the case of an accident, there is an increase in the reactivity of the reactor core. Consequently, with such a pressure drop, conventional reactor shut-down devices, such as the control rods, may be damaged to such an extent as to require replacement. Therefore, there has been a problem of quickly and reliably controlling the reactor activity when the pressure in the main coolant pipe line loop accidentally drops, and the object of the present invention is to solve this problem.